ML19262C556

Facility Cycle 2 Reload Rept. Revised by Util

ML19262C556
Person / Time
Site:	Davis Besse
Issue date:	02/01/1980
From:	BABCOCK & WILCOX CO.
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ML19262C555	List: ML19262C554 ML19262C556
References
BAW-1598, NUDOCS 8002140431
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Br.?-15 9 S January 1950 Revised ';y TECo February 6. 1980 DAVIS-SESSE NUCLEAR F0b'ER STATION UNIT 1, CYCLE 2 - RELOAD REPORT lMYEks r [j 3[l[7q [

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s r BAW-1598 January 1980 Revised by TECo Februa 1980

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7IS-BESSE NUCLEAR POWER STATION UNIT 1, CYCLE 2 - RELOAD REPORT r*

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BABCOCK & WILCOX Power Generation Group Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 Babcock & Wilcox

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CONTENTS Page

1. INTRODUCTION AND

SUMMARY

. .................... 1-1

2. OPERATING HISTORY ........................ 2-1

3. GENERAL DESCRIPTION .................... ... 3-1

4. FUEL SYSTEM DESIGN . ....................... 4-1 4.1. Fuel Assembly Mechanical Design . . . . . . . . . . .... 4-1 4.2. Fuel Red Design ...................... 4-1 4.2.1. Cladding Collapse . .. . . . . . . . . . . .... 4-1

4. 2. 2 '. Cladding Stress .. .. . . . . . . . .. . .... 4-1 4.2.3. Cladding Strain .. ... . . . . .. . . . .... 4-2 4.3. Thermal Design . ................... ... 4-2 4.4. Material Compatibility . . . ... . . . . . . . . . .... 4-2 4.5. Operating Experience . .............. . .... 4-2

5. NUCLEAR DESIGN . ..... ............... . .... 5-1 5.1. Physics Characteristics .. ... . . . . . . . . . .... 5-1 5.2. Changes in Nuclear Design . .. . . .. . . . . . . .... 5-2

6. THERMAL-HTDRAULIC DESIGN . ............... . .... 6-1

7. ACCIDENT AND TRANSIENT ANALYSIS . ... . . . . . . . . . .... 7-1 7.1. General Safety Analysis . . .. . . .. . . . .. . .... 7-1 7.2. Accident Evaluation ................. ... 7-1

8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS . . . . .... 8-1

9. STARTUP PROGRAM -- PHYSICS TESTING .. . ... ... .. . .... 9-1 9.1. Precritical Tests ..................... 9-1 9.1.1. Control Rod Trip Test ... . ... . .. . .... 9-1 9.1.2. Reactor Coolant Flow . .. . . . . . . . . . .... 9-1 9.1.3. RC Flow Coastdown . .. .. . . . . . . . . .... 9-1 9.2. Zero Power Physics Tests . . . .. . . . . . . .. . .... 9-2 9.2.1. Critical Boron Concentration . . .. . . . . .... 9-2 9.2.2. Temperature Reactivity Coefficient .. . . . .... 9-2 9.2.3. Control Rod Group Reactivity Worth . . .. . .... 9-2 9.2.4. Ejected Control Rod Reactivity Worth . . . . .... 9-3 9.3. Power Escalation Tests . . .... . . . . .. ... . .... 9-3 9.3.1. Core Power Distribution Verification at N40,rv75, and 'bl00% FP With Nominal Control Rod Position ... 9-3 Babcock t. Wilcox

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Page 9.3.2. Incore Vs Excore Detector Imbalance Correlation Verification at %40% FP . . . . . .... 9-5

! 9.3.3. Te=perature Reactivity Coefficient at %100% FP ... 9-5 I 9.3.4. Power Doppler Reactivity Coefficient at N100% FP .. 9-5 9.4. Procedure for Use When Acceptance Criteria Are Not Met ... 9-6

......................... ... A-1 f REFERENCES l

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r List of Tables Table 4-1. Fuel Design Parameters and Dimensions . . . . . . . . . .... 4-4 4-2. Fuel Thermal Analysis Parameters . . . . . . . . . . . . .... 4-5 5-3 5-1. Davis Besse 1 Cycle 2 Physics Parameters . . . . . . . . ....

5-2. Shutdown Margin Calculation for Davis Besse 1 Cycle 2 . .... 5-5

! 6-1. Cycles 1B and 2 Ther=al-Hydraulic Design Conditions --

Davis Besse . . . . . . . . . . .. . . .. . . . . . . .... 6-2 7-1. Comparison of Key Parameters for Accident Aanlysis . . .... 7-3 i 7-2. Bounding Values for Allowable LOCA Peak Linear Heat Rates ... 7-3 t

, List of Figures i Figut 3-1. Davis Besse 1, Cycle 2 Shuffle . .. . . .. . . . . . . .... 3-2 3-2. Enrichment and Burnup Distribution for Davis Besse 1, L Cycle 2 ....... . . . . . . . . . .. . . . . . . .... 3-3 3-3. Control Rod Locations for Davis Besse 1, Cycle 2 . . .. .... 3-4 5-1. BOC (4 EFPD), Cycle 2 Two-Dimensional Relative Power Distribution -- Full Power, Equilibrium Xenon,' APSRs Inserted . . 5-6 e

8-1. Reactor Core Safety Limit . . . . . . .. ... . . . . .... 8-19 8-2. Reactor Core Safety Limit . . . . .. ... . . . . . . .... 8-20 T- 8-21

8-3. Trip Setpoint for Flux-A Flux / Flow . .... . . . . . . ....

8-4. Allowable Value for Flux-A Flux / Flow . ;. . . . . . . . .... 8-22 8-5. Pressure / Temperature Limits . . . . . ... . . . . . . .... 8-23 8-6. Regulating Group Position Limits, O to 150 10 EFPD, Four

[. RCPs -- Davis Besse 1, Cycle 2 . . . . .. . . . . . . . .... 8-24 L.

8-7. Regulating Group Position Limits, Af ter 150 10 EFPD, Four RCPs -- Davis Besse 1, Cycle 2 . . .. ... , . . . . . .... 8-25

. 8-8. Regulating Group Position Limits, O to 150 10 EFPD, Three RCPs -- Davis Besse 1, Cycle 2 . . .. . .. . . . . . . .... 8-26

. 8-9. Regulating Group Position Limits, Af ter 150 10 EFPD, Three RCPs -- Davis Besse 1, Cycle 2 . . . . . . . . . . . . . .... 8-27 Babcock & Wilcox

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? t Figures (Cont'd)

Figure Page 8-10. APSR Position Limits, O to 150 10 EFPD, four RCPs - Davis-Besse 1, Cycle 2 . . ... . . . . . . . . . . . . . . . . . . ..... 8-28 8-11. APSR Position Limits, After 150 10 EFPD, Four RCPs --

Davis Besse 1, Cycle 2 . . . . . . . . . . . . . . . .... 8-29 8-12. APSR Position Limits, O to 150 10 EFPD, Three RCPs -

Davis Besse 1, Cycle 2 . . . . .. . . . . . . . ... .... 8-30 8-13. APSR Position L1=its, After 150 10 EFPD, Three RCPs --

Davis Besse 1, Cycle 2 . . . . . . . . . . . . . . . ..... 8-31 8-14. Axial Power I= balance Limits, O to 150 10 EFPD, Four RCPs -- Davis Besse 1, Cycle 2 . . . . . . . . . . .. ..... 8-32 8-15. Axial Power Imbalance Limits, After 150 10 EFPD, Four RCPs -- Davis Besse 1, Cycle 2 . . . . . . . . . . .. ..... 8-33 8-16. Axial Power Imbalance Limits, O to 150 10 EFPD, Three RCPs -- Davis Besse 1, Cycle 2 . . . . . . . . . . . . ..... 8-34 8-17. Axial Power Imbalance Limits, After 150 10 EFPD, Three RCPs -- Davis Besse 1, Cycle 2 . . . . . . . . . . . . ..... 8-35 8-18. Control Rod Core Locations and Group Assignments --

Davis-Besse 1, Cycle 2 . . . . . . . . . . . . . .. ..... 8-36

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1. INTRODUCTION AND

SUMMARY

This report justifies operation of the Davis Besse Nuclear Power Station Unit 1 at the rated core power of 2772 MWt for cycle 2. The required analyses are

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included as outlined in the USNRC document, " Guidance for Proposed License Amendments Relating to Refueling," June 1975. This report utilizes the analyt-ical techniques and design bases documented in several reports that have been submitted to the USNRC and approved by that agency.

Cycle 2 reactor and fuel parameters related to power capability are summarized in this report and compared to cycle 1. All accidents analyzed in the Davis Besse FSAR have been reviewed for cycle 2 operation and in cases where cycle 2 characteristics were conservative co= pared to cycle 1, no new analyses were performed.

Retainersl will be installed on the four fuel assemblies (FAs) containing neu-tron sources. The rerr.iners will provi;c positive retention during reactor operation. The effects on continued operation without orifice rod assemblies (ORAs) and the addition of retainers have been accounted for in the analysis performed for cycle 2.

The Technical Specifications have been reviewed and modified where required for cycle 2 operation. Based on the analyses performed, taking into account the ECCS Final Acceptance Criteria and postulated fuel densification effects, it is concluded that Davis Besse Unit 1, cycle 2 can be operated safely at its licensed core power level of 2772 MWt.

1-1 Babcock & Wilcox

2. OPERATING HISTORY The reference cycle for the nuclear and thermal-hydraulic analyses of Davis Besse Unit 1 is the operating cycle 1 which achieved criticality on August 12, 1977. On April 28,1978 af ter 86.7 effective full power days (EFPDs) of op-eration the plant was shutdown for removal of the burnable poison and orifice .

rod assemblies. In addition, eight FAs were relocated to reduce the flux shift towards the center of the core caused by the removal of the burnable poison rod asse=blies (BPRAs). After these modifications, criticality was achieved on July 23, 1978. No operating anomalies occured during cycle 1 op-eration that would adversely affect fuel performance during cycle 2.

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3. GENERAL DESCRIPTION

, The Davis Besse Unit 1 reactor core is described in detail in chapter 4 of the l

Final Safety Analysis Report 2 for the unit. The cycle 2 core consists of 177 I FAs, each of which is a 15 by 15 array containing 208 fuel rods, 16 control rod guide tubes, and one incore instrument guide tube. All FAs in batch 4 have a constant nominal fuel loading of 463.25 kg of uranium. Batches 1B, 2,

' and 3 have a fuel loading of 472.24 kg of uranium. The fuel consists of dished-end cylindrical pellets of uranium dioxide clad in cold-worked Zircaloy-4. The

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undensified nominal active fuel lengths, theoretical densities, fuel and fuel rod dimensions, and other related fuel parameters may be found in Tables 4-1 and 4-2 of this report.

Figure 3-1 is the core loading diagram for Davis Besse 1 cycle 2. The initial

,' enrichment of batch 1B is 1.98 wt % uranium-235. Batches 2, 3, and 4 have a 2.63, 2.96, and 3.04 uranium-235 enrichment, respectively. There will be 52 batch 1 assemblies discharged at the end of cycle 1; the remaining 4 batch 1 and all batch 2 and 3 assemblies will be shuffled to new locations. The batch 4 assemblies will occupy the periphery of the core. Note that the designation

' 1B is used to identify the batch 1 assemblies being reused for cycle 2. Batch 1A is now the remainder of batch 1 assemblies which have not been scheduled ,

for reinsertion. Figure 3-2 is an eighth-core map showing each assembly's burnup at the beginning-of-cycle (BOC) 2 and its initial enrichment.

Cycle 2 will be operated in a feed-and-bleed mode. The core reactivity con-trol will be supplied mainly by soluble boron and supplemented by 53 full

[ 1ength Ag-In-Cd control rod assemblies (CRAs). In addition to the full-length control rods, eight axial power shaping rods (APSRs) are provided for addi-

' The cycle 2 locations of the tional control of the axial power distribution.

61 control rods and the group designatione tre indicatea in Figure 3-3. Al-though the rod group designations differ, the core locations of the 61 control rods for cycle 2 are identical to those of the reference cycle 1.

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Figure 3 -l. Davis Besse 1, Cycle 2 Shuffle FUEL TRAKSFER CMAL )

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F F F F F A

F F F 2 3 2 F F F B F7 E8 F9 3 F 2 2 3 2 3 2 2 '

F 3 C 514 D5 C6 N3 C8 N13 C10 D11 F4 F F 18 3 2 2 3 2 2 D 3 15 F F C5 L1 E6 D7 Fil D9 E10 L15 E13 g F 2 3 2 3 3 2 3 3 2 3 2 F E4 A10 D4 L2 El E3 K15 Lie D12 A6 E12 F F 2 2 3 3 2 3 2 3 3 2 2 F F F F3 F5 a10 F12 37 F3 89 N2 86 Fl! F13 F 2 3 2 3 2 3 3 3 G

O 3 2 3 2 F C6 C12 G4 A9 C2 813 F8 43 C14 A7 C12 C4 C10 F 3 2 3 2 3 3 2 W-H IIS E3 E14 3 3 2 3 2 3' F .y ES M14 R14 88 E2 E2 H11 M2 E13 El F 2 3 2 3 2 3 3 3 2 3 2 3 2 F K6 812 K4 I 19 K2 C13 35 C3 K14 R7 K12 d4 K10 F F 2 2 3 3 2 l~ L3 L5 F10 3 2 3 3 2 2 F F D14 F7 811 F9 84 F6 til L13 F 2 3 2 3 3 2 3 3 M 2 3 2 F M R10 N4 F2 C1 MS CIS F14 N12 R6 M12 N F F IB 3 2 2 3 2 2 3 13 F F M3 F1 M6 N7 35 N9 M10 F15 fil w

3 F 2 2 3 2 3 2 2 0 F 3 312 N3 66 D3 ta D13 #10 ull D2 p F F F 2 3 2 F F F L7 A8 L9 F F F F F I

Z 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 XX Batch F Fresh Fuel YYY Previous Cycle Core heation 3-2 Babcock & Wilcox

Figure 3-2. Enrichment and Burnup Distribution for Davis Besse 1, Cycle 2 8 9 10 11 12 13 14 15 2.63 2.96 2.96 2.63 2.96 2.63 2.96 3.04 H

16,652 16,053 11,428 16,694 11,428 16,564 11,065 0 2.96 2.63 2.96 2.63 2.96 2.63 3.04 I K 8,974 14,960 10,547 16,168 13.430 16,383 0 2.96 2.96 2.63 2.63 3.04 3.04

_. L l 7,852 14,506 15,959 15,353 0 0

, 2.63 2.96 2.63 3.04 M

16,174 8,198 15,644 0 1.98 3.04 3.04 N

13,007 0 0 s

2.96 0

7,853 P

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Figure 3-3. Control Rod Locations for Davis Besse 1, Cycle 2 I

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B 4 7 4 C 1 6 6 1 D 7 8 2 8 7 E 1 5 5 1 1

F 4 8 7 5 7 8 l4 C 6 3 3 6 W- 7 2 5 3 5 2 7 -Y H

K 6 3 3 6 1'

l. 4 8 7 5 7 8 4 M 1 5 5 1 N 7 8 2 8 7 0 1 6 6 1 P 4 7 4 R

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1 2 3 4 5 6 -7 8 9 10 11 12 13 14 15 No. of Group rods Functions X Group Number 1 8 Safety 2 4 Safety 3 5 Safety 4 8 Safety 5 8 Control 6 8 Control 7 12 Control 8 8 APSRs Total # 61 3-4 Babcock & Wilcox

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4. FUEL SYSTEM DESIGN

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4.1. Fuel Assembly Mechanical Desien The cycle 2 core consists of the normal resident and reload Mark B fuel assem-

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blies (FAs). The pertinent fuel design parameters and dimensions are listed in Table 4-1. Retainer assemblies will be used on four FAs that contain the regenerative and primary neutron sources (RNS, PNS) . Justification for the

' design and use of the BPRA retainer assemblies is described in reference 1,

,_ which is applicable to these neutron source retainers. All FAs are identical in concept and are mechanically interchangeable.

4.2. Fuel Rod Design The fuel pellet end configuration has changed from a spherical dish for batches 1B through 3 to a truncated cone dish for batch 4. This mL or change facili-tates manufacturing while =aintaining the same end void factor. Fuel pellet density for batch 4 is 94% TD compared to 96% for batches 1B, 2, and 3. The mechanical evaluation of the fuel rod is discussed below.

,_ 4.2.1. Cladding Collapse I

Creep collapse analyses were performed for two cycle assembly power histories for Davis Besse 1. Batches 1B, 2, and 3 are more limiting than batch 4 due u to their previous incore exposure time. A ba*.ch 2 FA was determined to have the most limiting power history and was, ther : fore, analyzed for creep col-lapse.

The limiting power history was used to calculate the fast neutron flux level for the energy range >l MeV. The collapse time for the most limiting assembly was cor vntively determined to be >30,000 ef fective full power hours (EFPHs),

which is *er than the maximum three cycle design life (Table 4-1).

This analysis was performeo based on ti onditions set forth in reference 3.

4.2.2. Cladding Stress The Davis Besse I stress parameters are enveloped by a conservative fuel rod stress analysis. For design evaluation the primary membrane stress must be 4-1 Babcock & kVilcox

1ess than two-thirds of the minimum specified unirradiated yield strengch and all stresses must be less than th'e 'caxi.num specified unirradiated yield strength. In all cases, the =argin is in excess of 30%. The following con-servatisms with respect to Davis Besse 1 fuel were used in the analysis: ,

1. A lower post-densification internal pressure.

2. A lower initial pellet density.

3. A higher system pressure.

4. A higher thermal gradient across the cladding.

4.2.3. Claddi g Strain The fuel design criteria specify a limit of 1.0% on cladding tensile circum-ferential plastic strain. The pellet design is so that the tensile plasti cladding strain is less than 1% at 55,000 mwd /mtU. The following cladding strain conservatisms are applicable with respect to the Davis Besse 1 fuel:

1. The maximum specification value for the fuel pellet diameter was used.

2. The maximum specification value for the fuet density was used.

3. The cladding inside diameter used was the icwest permitted specification tolerance.

4. The maximum expected three cycle local pellet burnup is less than 55,000 mwd /mtU.

4.3. Thermal Design All FAs in the Davis Besse cycle 2 core are thermally similar. The fuel pa-rameters, desigu linear heat rar.e (LHR) capacity, and average fuel temperature for each fuel batch in cycle 2 are given in Table 4.2. LHR capa-bilities are based on centerline fuel melt and verr 9stablished by use of the TAFY-3 code 4 with fuel densification to 96.5% theoretical density. Re-sinter data for batch 4 has been checked for fuel melt limit and has been found acceptable.

4.4. Material Compatibility The chemical compatibility of all possible fuel cladding-coolant assembly interactions for the batch 4 FAs is identical to that of the present fuel.

4.5. Operating Experience Babcock & Wilcox operating experience with the Mark B 15 by 15 FA has verified the adequacy of its design. As of September 30, 1979, the following experience 4-2 Babcock & VVilcox

has been accumulated for the eight operating B&W 177-FA plants using the Mark B FA:

Maximum assembir( } """ * "* "* (b) burnup , mwd /mtU Current -

electrical output, Reactor cycle Inc. ore Discharged FSat Oconee 1 6 18,610 40,000 28,291.915 l Oconee 2 4 27,900 33,700 23,829,671

.- Oconee 3 5 23,100 29,400 23,492,274 TMI-l 4 32.400 32,200 23,857,504 ANO-1 4 23,600 33,222 21,462,382

! Rar::ho Seco 3 37,462 29,378 18,400,062 Crystal River 3 2 20,656 15,264 8,718,554 Davis Besse 1 1 11,600 -- 5,755,467

(*)As of December 31, 1979.

(b)As of September 30, 1979.

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Tabic 4-1. Fuel Design Parameters and Dimer- ,jn s Batch 1B Batch 2 Batch 3 Batch 4 Fuel assembly type Mark B-4A Mark B-4A Mark B-4A Mark B-4A No. of assemblies 4 61 60 52 Fuel rod OD, in. 0.430 0.430 0.430 0.430 Fuel rod ID, in. 0.377 0.377 0.377 0.377 Flexible spacers, type Spring Spring Spring Spring Rigid spacers, type Zr-4 Zr-4 Zr-4 Zr-4 Undensified active fuel length, in. 143 5 143.5 143.5 143.44 Fuel pellet OD (mean specified), in. 0.3675 0.3675 0.3675 0.3697 Mean specified fuel pellet initial density, % TD 96.0 96.0 96.0 94.0

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L Initial fuel enrichment, wt % 235U l.98 2.63 2.96 3.04 Estimated residence (max), EFPIl 18,456 18,456 25,656 21,216 Cladding collapse time, EFPII >30,000 >30,000 >30,000 >30,000

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Table 4-2. Fuel Thermal Analysis Parameters Batch 1B Batch 2 Batch 3 Batch 4 No. of assemblies 4 61 60 52

Nominal pellet density, % TD 96 96 96 94 Pellet diameter, in. 0.3675 0.3675 0.3675 0.3697 Stack height, in. 143.0 143.5 143.5 143.44 Densified Fuel Parameters (*

I Pellet diameter, in. 0.3651 0.3651 0.3651 0.3648 Fuel stack he.c n in. 143.14 143.14 143.14 141.65 Average LHR @ 2772 MWt, kW/ft 6.14 6.14 6.14 6.21 Fuel T at nominal LHR, *F 1340 1340 1340 1355

,! LHR to [ fuel melt, kW/ft 20.4(D) 20.4 20. 4 (b) 20.4

- Note: Core average densified LHR at 2772 MWt is 6.16 kW/ft.

j (*)Densification to 96.5% TD is assumed.

E ( )Two batch 1D and five batch 3 YAs have LHR to centerline nelt limits of 20.17 and 20.35 kW/ft, respectively. All FAs in batches 1B and 3 are limited to less than 20.17 kW/ft by RPS setpoints.

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5. NUCLEAR DESIGN 5.1. Physics Characteristics Table 5-1 coepares the core physics parameters of cycles 1 and 2. These val-ues were generated using PDQ075 for both cycles. Since the core has not yet reached an equilibrium cycle, differences in core physics parameters are to be expected between the cycles. Figure 5-1 illustrates a representative rel-ative power distribution for BOC 2 at full power (FP) with equilibrium xenon and group 8 inserted.

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The critical boron concentrations for cycle 2 are lower than those of the ref-erence cycle 1 due to the difference in design cycle lengths. The hot full l power (HFP) control rod worths are different because in cycle 2 the bank loca-tic as and designations have changed from those of the reference cycle. Control rod worths are sufficient to maintain the required shutdown margin as indicated in Table 5-2. The ejected rod worths in Table 5-1 are the maximum calculated I values. It is difficult to compare maximum ejected tod worths between cycles since neither the rod patterns from which the rod is assumed ejected nor the isotopic distributions are identical. Calculated ejected rod w rths and their adherence to criteria are considered at all times in life and at all power

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levels in the development of the rod position limits presented in section 8.

The maximum stuck rod worths for cycle 2 are also different than those for cycle 1. The adequacy of the shutdown margin with cycle 2 stuck rod worths is shown in Table 5-2. The following conservatisms were applied for the shut-

, down calculations:

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1. Poison material depletion allowance.

2. 10% uncertainty on net rod worth.

3. Flux redistribution penalty.

Flux redistribution was taken into account since the shutdown analysis was cal-culated using a two-dimensional model. The cycle 2 power deficits from HZP to HFP are higher than those for cycle 1 due to the more negative moderator

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5-1 Babcock & \Vilcox

coefficients in cycle 2. The differential boron and xenon worths are different ,

at B0C because the reference (first) cycle had no plutonium present. In cycle 2, the presence of plutonium causes spectrum shifts which results in differences in these parameters. At EOC the differential boron and xenon worths are simi-lar in both cycles. The effective delayed neutron fractions for cycle 2 show a decrease with burnup (also reported in the reference cycle 1).

5.2. Changes in Nuclear Design There are no significant core design changes between the reference and the cycle 2 designs. The same calculational methods and design information were used to obtain the important nuclear design parameters. No significant opera-tional or procedural changes exist with regard to axial or radial power shape, xenon, or tilt control. The operational and RPS limits CIechnical Specifica-tion changes) for cycle 2 are presented in section 8.

5-2 Babcock & Wilcox

Table 5-1. Davis Besse 1 Cycle 2 Physics Parameters (*)

Cycle 1 Cycle 2 Cycle length, EFPD 433 274 Cycle burnup, mwd /mtU 14,360 9,108 Average core burnup -- EOC, mwd /mtU 14,360 18,680 Initial core loading, mtU 83.6 83.4 Critical boron -- BOC, ppm (no Xe)

HZP(b), group 8 (37.5% wd) 1,520 1,360 HFP, group 8 inserted 1,408 1,167 Critical boron - EOC, ppm (eg Xe) p group 8 (37.5% wd, eq Xe) ff9 2 Control rod worths - HFP, BOC, % Ak/k

, Group 6 2.00 0.79 Group 7 1.50 1.52 Group 8 (37.5% vd) 0.46 0.30 Control rod worths - HFP, EOC, % Ak/k Group 7 1.12 1.60 Group 8 (37.5% wd) 0.30 0.39 Max ejected rod worth - HZP, % Ak/k("}

BOC 0.85 0.69 EOC 0.63 0.73 Max stuck rod worth - HZP, % Ak/k 2.98 1.60 BOC EOC 1.25 1.51

' Power deficit, HZP to HFP, % Ak/k BOC -1.00 -1.61 EOC -2.24 -2.34 Doppler coeff -- BOC 10-5 (Ak/k/*F) 100% power (no Xe) BOC -1.25 -1.53 100% power (eq Xe) EOC -1.45 -1.59 Moderator coeff - HFP,10-4 (Ak/k/*F)

BOC (no Xe, 1064 ppm, group 8 in) -0.06 -0.69 EOC (eq Xe, 17 ppm, group 8 in) -2.60 -2.84

. Boron worth - HFP, ppm /% Ak/k BOC (1150 ppm) 99 111 E0C (17 ppm) 102 97 Xenon worth -- HFP, % Ak/k BOC (4 EFPD) 2.73 2.61 EOC (equilibri~um) 2.71 2.78 5-3 Babcock & kVilcox

Table 5-1. (Cont'd)

Cycle 1 Cycle 2 Effective delayed neutron fraction - HFP BOC

• 0.00690 0.00586 EOC 0.00514 0.00544

(* Table 5-1 contains the cycle 1 values that are for the original 433 EFPD cycle 1 design.2 The modified cycle 1 had a design cycle length of 485 EFPD, however, it is now planned to refuel after 414 EFPD. The 414 FJPD cycle 1 length has been input to the cycle 2 calculations as presented in this report.

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HZP denotes hot zero power (532F T##E); HFP denotes hot full power (584F core Tavg).

( } Ejected rod worth for groups 5 through 8 inserted.

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Table 5-2. Shutdown Margin Calculation for Davis Besse 1 Cycle 2 BOC, EOC,

% Ak/k  % ak/k Available Rod Worth Total rod worth, HZP(*) 7.28 7.67 Worth reduction due to burnup of z poison material -0.03 -0.03 Maximum stuck rod, HZP -1.60 -1.51 Net worth 5.65 6.13 Less 10% uncertainty -0.57 -0.61 Total available worth 5.08 5.52 Required Rod Worth Power deficit, RFP to HZP 1.61 2.34 Max allowable inserted rod worth 0.26 0.39 Flux redistribution 0.60 1.15 Total required worth 2.47 3.88 Shutdown Margin Total available minus total required 2.61 1.64 Note: Required shutdown margin is 1.00% ok/k (a)RZP denotes hot zero power (532F T""E); HFP denotes hot full power (584F core T ).

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. . I, Figure 5-1. BOC (4 EFPD), Cycle 2 Two-Dicensional Relative Power Distribution -- Full Power, Equilibrium _

Xenon, APSRs Inserted (a) 13 14 15 10 11 12 8 9 0.88 1.00 0.89 1.25 1.22 0.99 1.01 H 1.14 .-

0.89 0.97 0.92 0.87 1.33 1.14 1.13 K

8 0.80 0.93 1.22 0.74 1.27 1.06 L

H 0.97 1.10 0.98 1.05 0.93 1.21 0.76 N

0.73 0

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X Inserted Rod Croup Nu=b'er X.XX Relative Power Density b") Calculated results from two-dimensional pin-by-pin PDQ07.5 Babcock & Wilcox 5-6

6. THERMAL-HYDRAULIC DESIGN The incoming batch 4 fuel is hydraulically and geometrically similar to the fuel remaining in the core from cycle 1. The thermal-hydraulic design evalua-tion supporting cycle 2 operations used the models and methods d~ :ribed in references 1, 6, and 7.

The flux / flow trip setpoint of 1.07 has been established for cycle 2 operation.

This setpoint and other plant operating limits based on DNBR criteria contain margin to the design minimum DNBR of 1.30.B&W-2 to account for the DNB rod bow penalty.

A rod bow penalty has been calculated according to the procedure approved in reference 8. The burnup used is based on the maximum FA burnup of the batch that contains the FA with the maximum radial x local peak. For cycle 2, this burnup is 26,654 mwd /mtU in a batch 3 assembly. The resulting net rod bow penalty is 1.8% af ter the 1% flow area reduction factor credit is included.

The rod bow. penalty is accounted for by including DNBR margin in trip set-points and operating limits.

4 6-1 Babcock & \%Icox

Table 6-1. Cycles IB(b) and 2 Thermal-Hydraulic Design Conditions -- Davis Besse Cycle 1B Cycle 2 Design power level, MWt 2772 2772 System pressure, psia 2200 2200 Reactor coolant flow, % design 110 110 Vessel inlet / outlet coolant temp.,

100% power, F 557.7/606.3 557.7/606.3 Ref design radial-local power peaking factor , 1.71 1.71 Ref design axial flux shape 1.5 cosine 1.5 cosine with tails with tails Hot channel factors Enthalpy rise.(F ) , 1.011 . 1.011 S

Heat flux (F")

9 1.014 1.014 Flow area 0.98 0.98 Active fuel length See Table 4-2 See Table 4-2 Avg heat flux, 100% power, Btu /h-ft2 1.86 x 105 (a) 3.89 x 105 (a)

Max heat flux, 100% power, Btu /h-ft2 4.78 x 105 (a) 4.83 x 105 (a)

CHF correlation BAW-2 BAW-2 Minicus DNBR, (% power) 1.81 (112%) 1.79 (112%)

(*)With thermally expanded fuel rod OD of 0.43075 inch.

(.b)After removal of burnable poison and orifice rod asse=blies.

6-2 Babcock r. Wilcox

r 7. ACCIDENT AND TRANSIENT ANALYSIS

{

I 7.1. General Safety Analysis

(

Each FSAR2 accident analysis has been examined with respect to changes in cy-F cle 2 parameters to determine the effect of the cycle 2 reload and to ensure

{

that thermal performance during hypoth'etical transients is not degraded. The effects of fuel densification on the FSAR accident results have been evaluated and are reported in reference 7.

f Improved fuel utilization and the inherent increase in core average burnup experienced in cycle 2 have resulted in a higher plutonium-to-uranium fission ratio than that used in the FSAR. A study of the major FSAR chapter 15 acci-dents using the cycle 2 iodine and noble gas inventories concluded that the thyroid and whole body doses were well below the 10 CFR 100 limits.

7.2. Accident Evaluation The key parameters that have the greatest effect on determining the outcome of a transient can typically be classified in three major areas: (1) core thermal, (2) thermal-hydraulic, and (3) kinetics parameters including the re-

[

activity feedback coefficients and control rod worths.

Fuel thermal analysis parameters from each batch in cycle 2 are given in Table 4-2. A comparison or the cyc e. 2 thermal-hydraulic maximum design conditions to the previous cycle values is presented in Table 6-1. A comparison of the i key kinetics parameters from the FSAR and cycle 2 is provided in Table 7-1.

A generic LOCA analysis for B&W 177-FA raised-loop NSS has been performed us-

' ing the Final Acceptance Criteria ECCS Evaluation Model.9 The combination of average fuel temperature as a function of LER and the lifetima pin pressure v data used in the LOCA limits analysis is conservative compared to those cal-culated for this reload. Thus, the analysis and the LOCA limits reported in reference 9 provide conservative results for the operation of Davis Besse 1 cycle 2 fuel. A tabulation showing the bounding values for allowable LOCA peak linear heat rates for Davis Besse 1 cycle 2 fuel are provided in Table 7-2.

7-1 Babcock & Wilcox

lic, and of cycle 2 core thermal, thermal-hydrau this It is concluded by examinatiot. l revious cycle values that kinetics properties with respect to hacceptab e pbility to safely operate the Da basis core reload vill not adversely af f ect t e aConsidering the previously 2 is Besse 1 plant during cycle 2. the transient evaluation of cycle used in the FSAR and subsequentl cycles, accepted analyses. The initial condi-onsidered to be bounded by previous y bounded by the FSAR and/or the fuel tions of the transients in cycle 2 are 7

densification report Babcock & Wilcox 7-2

Table 7-1. Comparison 'f Key Parameters for Accident Analysis FSAR, Predicted densif cycle 2 r_ Parameter value value BOL Doppler coeff, 10-5 Ak/k/*F -1.28 -1.53 i- EOL Doppler coeff, 10-5, Ak/k/*F -1.45(*) -1.59

( BOL moderator coeff, 10-4, ok/k/*F +0.13 -0.69 EOL moderator cceff,10-4, ok/k/*F -3.0 -2.84 f All rod bank worth (HZP), % ak/k 10.0 7.28 Boron reactivity worth (HFP), ppm /1% ok/k 100 ill r

Max ejected rod worth (EFP), % Ak/k 0.65 0.36 t

Max dropped rod worth (EFP), % Ak/k 0.65 0.20 Initial boron cone (HFP), ppm 1407 1170

(*) 1.77 x 10-5 ok/k/*F was used for steamline failure analysis.

Table 7-2. Bounding Values for Allowable i LOCA Peak Linear Heat Rates

-- Allowable

Core peak linear elevation, heat rate, ft kW/ft t 2 16.5

, 4 17.2 t 6 18.4 8 17.5 ft- 10 17.0

(. .

9 7-3 Babcock & \Vilcox

8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS f

The Technical Specifications have been revised for cycle 2 operation to ac-count for changes in power peaking and control rod worths. In addition, changes were the result of the following:

1. The reduction of the fuel rod bow DNB penalty8 has permitted the relaxa-

[ tion of certain operating limits.

(

2. The required quantity and concentration of boric acid necessary to reach a cold shutdown condition have been reviewed and found to be within pre-vious Technical Specification limits. Page B 3/4 1-2 of the Technical

~

Specifications has been changed so that the basis is consistent with cycle 2 requirements.

3. All references to two-pump operation have been removed to avoid confusion since two-pump operation is not allowed by the operating license. Figures

' 3.1-3c, 3.1-3d, 3.2-3a, and 3.2-3b have been removed.

4. The DNB limits in the bases have been changed from 1.32 to 1.30 to be

{ consistent with the approved limit of the B&W-2 correlation.

5. Sections 3.1.3.9 and 4.1.3.9 have been aoded to provide axia] power shap-f ing rod group limits for physical insertion.

t

6. The high pressure limits appearing in the Technical Specifications, Table 1 2.2-1, has been previously submitted, but not acted on, L'

in a letter from Toledo Edison Company to NRC, Serial Number 527, July

[ 13, 1979, see attachment 1 to this report.

i Based on the Technical Specifications derived from the analyses presented in

[

l this report, the Final Acceptance Criteria ECCS limits will not be exceeded, nor will the thermal design criteria be violated.

8-1 Babcock & Wilcox

Table 2.2-1. Reactor Protection System Instrumentation Trip Setpoints Functional unit Trip setpoint Allowable values

1. Panual reactor trip Not applicable. Not applicable.

2. liigh flux 5105.5% of rated thermal power with $105.6% of rated thermal power with four pumps operating four pumps operatingI

$80.2% of rated thermal power with s80.3% of rated thermal power with three pumps operating three pumps operating #

3. RC high temperature s619'F s619.08'F

4. Flux - A flux / flow ( } Trip setpoint not to exceed the lim- Allowable values not to exceed the it line of Figure 2.2-1 limit line of Figure 2.2-2 I

5. RC low pressure ( } 21985 psig 21984.0 psig* 21976.5 psig**

?$

w

6. RC high pressure s2300 psig s2301.0 psig* s2308.5 psig**

7. RC pressure-temperature ( } 2(12.60 T F - 5660) psig 2(12.60 T, *F - 5660.41) psig#

out lT 8

x N

a-

2.1. SAFETY LIMITS BASES 2.1.1 and 2.1.2 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation te=perature.

Operation above the upper boundary of the nucleate boiling regime would re-sult in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat trans-fer coefficient. DNB is not a directly measurable parameter during operation and therefore thermal power and reactor coolant temperature and pressure have been related to DNB through the B&W-2 DNB correlation. The DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location'to the local heat flux, is indicative of the =argin to DNB.

The minimum value of the DNBR during steady state operation, normal operation-al transients, and anticipated transients is limited to 1.30. This value cor-responds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curve presented in Figure 2.1-1 represents the conditions at which a mini-mum DNBR of 1.30 is predicted for the maximum possible thermal power 112% when the reactor coolant flow is 387, 200 gpm, which is 110% of design flow rate for four ope. rating reactor coolant pumps. This curve is based on the follow-ing hot channel f actors with potential fuel densification and fuel rod bowing effects:

F = 2.56; F "

q H"

• I

• The design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to minimum al-lowable control rod withdrawal, and form the core DNBR design basis.

DAVIS-BESSE, UNIT 1 B 2-1 8-3

l SAFETY LIMITS BASES The reactor trip envelope appears to approach the safety limit more closely than it actually does because the reactor trip pressures are measured at a location where the indicated pressure is about 30 psi less than core outlet pressure, providing a more conservative margin to the safety limit.

The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and account for the effects of potential fuel densification and poten-tial fuel red bow:

1.

The 2.56 1.30 DNBR or the limit produced combination of the by a nuclear radial peak,power axial peaking factor peak, and of Fq o=f position the axial peak that yields no less than a 1.30 DNBR.

2. The combinatica of radial and axiu. peak that causes central fuel melting at the hot spot. The limit is 20.4 kW/ft.

Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalance produced by the power peaking.

~

The specified flow rates for curves 1 and 2 of Eigure 2.1-2 correspond to the cxpected minimum flow rates with four pumps and three pumps, respectively.

The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-maximum ther=al power combinations shown in bases Figure 2.1.

The curve of bases, Figure 2.1 represent the conditions at which a minimum DNBR of 1.30 is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation or the local quality at the point of mini-mum DNER is equal to +22%, whichever condition is more restrictive. This curve includes the potential effects of fuel rod bow and fuel densification.

The DNBR as calculated by the 3&W-2 DNB correlation continually increases from pctnt of minimum DNBR, so that the exit DNBR is alv ys higher. Extrapolation of the correlation beyond its published quality range of +22% is justified on the basis of experi= ental data.

DAVIS-BESSE, UNIT 1 B 2-2 3-4

SAFETY LIMITS BASES

'~

For the curve of bases Figure 2.1, a pressure-te=perature point above and to the lef t of the curve would result in a DNBR greater than 1.30 or a local l quality ac the point of minimum DNER less than +22% for that particular reac-tor coolant pump situation. The 1.30 DNBR curve for four pu=p operation is l more restrictive than any other reactor coolant pump situation because any pressure /teeperature point above ar.d to the left of the four pu=p curve will be above and to the left of the three pu=p cu rve. l 2.1.3. REACTOR COOLANT SYSIEM PRESSURE The restriction of this safety limit protects the integrity of the reactor cool-ant system from overpreesurization and thereby prevents the release of radio-nuclides contained in t..e reactor coolar,t from reaching the containment atmos-phere. -

• The reactor pressure vessel and pressurizer are designed to Sec~'on III of the ASME Boiler and Pressure Vessel Code which permits a maximum tratisient pres-sure of 110%, 2750 psig, of design pressure. The reactor coolant system pip-ing, valves and fittings, are designed to ANSI B 31.7, 1968 Edition, which permits a maximum transient pressure of 110%, 2750 psig, of component design pressure. The safety limit of 2750 psig is therefore consistent with the de-sign criteria and associated code require =ents.

The entirt reactor coolant system is hydrotested at 3125 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.

DAVIS-BESSE, UNIT 1 B 2-3 8-5 -

LIMIIING SAFETY SYSTEM SETTINGS BASES RC High Tenperature The RC high temperature trip 1619'F prevents the reactor outlet temperature from exceeding the design limits and acts as a backup trip for all power ex-cursion transients.

Flux -- A Flux / Flow "he power level trip setpoint produced by the reactor coolant system flow is based on a flux-to-flow ratio which has been established to accommodate flow decreasing transients from high power where protection is not provided by the high flux / number of reactor coolant pumps on trips.

The power icvel trip setpoint produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rata decreases. The power level setpoint produced by the power-to-flow ratio provides overpower DNB protection for all modes of pu=p operation. For every flow rate there is a maximum per-missible power level, and for every power level there is a minimum permissible lw, ilev rate. Examples of typical power level _and low flow rate co=Linations for the pump situations of Table 2.2-1 that would result in a trip are as follows:

1. Trip would occur when four reactor coolant pumps are ope. rating if power is 107% and reactor coolant flow rate is 100% of full flow rate, or flow rate is 93.3% of full flow rate and power level is 100%.

2. Trip would occur when three reactor coolant pumps are operating if power is 80.2% and reactor coolant flow rate is 74.9% of full flow rate, or flow rate is 69.8% of full flow rate and power is 75%.

For safety calculations the maximum calibration and instrumentation errors for the power level were used. Full flow rate in the above two examples is defined as the flow calculated by the heat balance at 100% power.

1 DAVIS-BESSE, UNIT 1 B 2-5 8-6

LIMITING SAFFTY SYSTEM SETTINGS

~

BASES i The axial power imbalance boundaries are established in order to prevent re-actor thermal limits fron being exceeded. These thermal limits are either power peaking kW/f t limits or DNBR limits. The axial power imbalance reduces the power level trip produced by a flux-co-flow ratio such that the boundaries of Figure 2.2-1 are produced.

f I

t RC Pressure - Low, High, and Pressure Temperature

' The high and low trips are provided to limit the pressure range 1. shich re-

, actor operation is permitted.

r- During a slow reactivity insertion startup accident from low power or a slow j reactivity insertion from high power, the RC high pressure setpoint is reached before the high flux trip setpoint. The trip setpoint for RC high pressure, 2300 psig, has been established to maintain the system pressure below the safe-ty limit, 2750 psig, for any design transient. The RC high pressur9 trip is bucked up by the pressurf. er code safety valves for RCS over pressure protec-tion, and is therefore set lower than the set pressure for these valves, 2435 psig. The RC high pressure trip also backs up the high flux trip.

The RC low pressure, 1983 psig, and RC pressure-temperature (12.60 Tout -

5660) psig, trip setpoints have been established to maintain the DNB ratio I greater than or equal to 1.30 for those design accidents that result in a pressure reduction. It also prevents reactor operation at pressures below the valid range of DNB correlation limits, protecting against DNB.

High Flux / Number of Reactor Coolant Pumps On

[ In conjunction with the flux - A flux / flow trip the high flux / number of reac-tor coolant pumps on trip prevents the mini =um core DNBR from decreasing below 1.30 by tripping the reactor due to the loss of reactor coolant pump (s). The l pump monitors also restrict the power level for the number of pumps in opera-tion.

t b

m DAVIS-BESSE, UNIT 1 B 2-6 8-7

l REACTIVITY CONTROL SYSTEMS REGULATING ROD INSERTION LIMITS _

LIMITING CONDITION FOR OPERATION l 3.1.3.6 The regulating rod groups shall be limited in physical insertion as shown on Figures 3.1-2a and -2b and 3.1-3a and -3b, with a rod group overlap of 25 5% between sequential withdrawn groups 5, 6, and 7.

APPLICABILITY: MODES 1* and 2*#,

ACTION:

(in With the_ regulating rod _ groups inserted beyond the above insertion limit,s a region _ other than a.cceptable operation), or with any_ group _ sequence.or _over-

lap out_ side the speci_fied limits, .except for. surveillance test _ing pursuant._to__ .

Specification 4.1.3.1.2, either: _ _ _

a. Restore the regulating groups to within the limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or

b. Reduce thermal power to less than or equal to that fraction of rated ther-cal power which is allowed by the rod group position using the above fig-ures within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or

c. Be in at least hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

NOTE: If in unacceptable region, also cee Section 3/4.1.1.1.

• Sca Special Test Exceptions 3,10.1 and 3.10.2.

1. With k,ff 1 1.0.

DAVIS-BESSE, UNIT 1 3/4 1-26

~ a'>

8-8

lREACTIVITY CONT 90L SYSTEMS AXIAI D' WER SHAP13C .0D INSERTION LIMITS

. LIMIT' ONDITION FOR OPERATION 9

3.1.3.9 The axial power shaping rod group shall be limited in physical inser-tion as shown on Figures 3.1-Sa, -Sb, -Sc, and -5d.

APPLICABILITY: MODES 1 and 2*.

~~

ACTION:

With the axial power shaping rod group outside the above insertion limits,

, - either:

a. Restore the axial power shaping rod group to within the limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or l b. Reduce thermal power to less than or equal to that fraction of rated ther-mal power which is allowed by the rod group position using the above fig-ures within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or

c. Be in at least hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS L

l 1- 4.1.3.9 The position of the axial power shaping rod group shcIl be deter-

' mined to be within the insertion limits at least once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

f i

• With k,gg 2 1.0.

3/4 1-34

3/4.2. POWER DISTRIBUTION LIMITS AXIAL POW 1.1 IMBAIMCE LIMITING CONDITION FOR OPERATION 3.2.1 Axial power imbalance shall be maintained within the limits shown on Figures 3.2-1 and 3.2-2.

APPLICABILITY: MODE 1 above 40% of rated thermal power

• ACTION:

With axial power imbalance exceeding the limits specified above, either;

a. Restore the axial power imbalance to within its limits within 15 minutes, el

b. Reduce power _until imbalance limits are met. __ ___ __. . . . - . _. _ __

+

SURVEILLANCE REQUIREMENTS 4.2.1 The axial power imbalance shall be determined to be within limits at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when above 40% of rated thermal power except when the axial power imbalance alarm is inoperable, then calculate the axial power imbalance at least once per hour.

• See Special Test Exception 3.10.1.

DAVIS-BESSE, UNIT 1 3/4 2-1 8-10

P0k'ER DISTRIBUTION LIMITS DNB PARAMETERS

6. ' LIMITING CONDITION FOR OPERATION r-3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1:

a. Reactor coolant hot leg temperature.

b. Reactor coolant pressure.

,, c. Reactor coolant flow rate.

APPLICABILITY: MODE 1.

a.

ACTI0tJ:

. If parameter a or b exceeds its limit, restore the parameter to within its l'mit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce thermal power to less than 5% of raf.ed thermal power within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

If parameter c exceeds its limit, eithe r:

1. Restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or

2. Limit thermal power at least 2% below rated thermal power for each 1% para-meter e is outside its limit for four pump operation within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or limit thermal power at least 2% below 75%.of. rated thermal power for each 1% parameter e is outside its limit for 3 pump operation within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

. ~

SURVETLLANCE REOUIREMENTS

. 4.2.5.1 Each of the parameters of Table- 3.2-l__shd11_be verified to be within their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. ,

4.2.5.2 The reactor coolant system total flow rate shall be determined to be within its limit by measurement at least once per 18 months.

.~ _

DAVIS-3 ESSE, UNIT 1 3/4 2-13 8-11

Table 3.2-1. DNB Margin Limits Four RC Three RC pu=ps pumps Parameter operating operating RC hot leg te=p.,

T'H RC pressure, psig( } 22062.7 22058.7(*)

RC flow rate, gpm(c) 2396,880 2297,340

(*) Applicable to the loop with two RC pumps

. operating.

( )Lilit not applicable duEing either a ther-mal power ramp increase in excess of 5% of rated ther=al power per minute or a thermal power step increase of greater than 10% of rated thermal power.

(")These flows include a flow rate uncertainty of 2.5%. .

DAVIS-BESSE, UNIT 1 3/4 2-14 8-12

3/4.4. REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION

~

3.4.1 Both reactor coolant loops and both reactor coolant pumps in each loop shall be in operation.

APPLICABILITY: As noted below, but excluding MODE 6.*

ACTION:

l MODES 1 and 2:

a. With one reactor coolant pump not in operation, startup and power opera-J' tion may be initiated and may proceed provided thereal power is restricted r to less than 80.2% of rated ther=al power and within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the setpoints l

. for the following trips have been reduced to the values specified in Spec-ification 2.2.1 for operation with three reactor coolant pumps operating:

1. High flux

2. Flux-a flux / flow.

I

• See Special Test Exception 3.10.3, i

DAVIS-BESSE, UNIT 1 3/4 4-1 8-13

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.4. MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the reactor coolant system average temperature less than 525*F. This limitation is required to ensure (1) the moderator temperature coef ficient is within its analyzed te=perature range, (2) the protective instrumentation is within its nor=al operating range, (3) the pressurizer is capable of being in an operable status with a steam bubble, and (4) the reactor pressure vessel is above its minimum RTET temperature.

2/4.1.2. BORATION SYSTEMS The boron injection system ensures that negative reactivity control is avail-able during each mode of facility operation. The components required to per-form this furecion include (1) borated water sources, (2) makeup or DHR pu=ps, (3) separate flow paths, (4) boric acid pu=ps, (5) associated Seat tracing systems, and (6) nn emergency power supply from operable emergency busses.

With the RCS average temperature above 200*F, a minimum of two separate and redundant boron injection syste=s are provided to ensure single functional capability in the event an assumed failure renders one of the systems inop-erable. Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period.

The boration capability of either system is sufficient to provide a shutdown margin from all operating conditions of 1.0% ak/k af ter xenon decay and cool-down to 200*F. The maxi =um boration capability requirement occurs from full l power equilibrium xenon conditions and requires the equivalent of either 7373 gallons of 8742 ppm borated water from the boric acid storage tanks or 52,726 gallons of 1800 ppm borated water from the borated water storage tank.

The requirements for a minimum contained volume of 434,650 gallons of borated water in the borated water storage tank ensures the capability for borating the RCS to the desired level. The specified quantity of borated water is cca-sistent with the ECCS requirements of Specification 3.5.4, therefore, the larger volume of borated water is specified.

With the RCS. temperature below 200*F, one injection system is acceptable with-out single failure consideration on the basis of the DAVIS-BESSE, UNIT 1 B 3/4 1-2 8-14

REACTIVITY CONTROL SYSTDiS BASEE 3/4.1.3. MOVABLE CONTROL ASSEMBLIES (Continued)

The maximum rod drop time permitted is consistent with the assumed rod drop time used in the safety analyses. Measurement with Tavg 2 525*F and with re-actor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operat-ing conditicns.

Control rod positions and operability of the rod position indi.ttors are re-quired to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with frequent verifications required if an automatic monitorir.g channel is inoperable. These verification frequencies are rdequate for assuring that the applicable LCO's are satisfied.

The limitation on thermal power based on xenon reactivity is necessary to en-sure that power peaking limits are not exceeded even with specified rod in-sertion itsits satisfied.

The li=itation on axial power shaping rod insertion is necessary to ensure that power peaking limits are not exceeded.

4 DAVIS-BESSE, UNIT 1 B 3/4 1-4 8-15

. . I 3/4.2. POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during -

Condition I (normal operation) and II (incidents of moderate frequency) events ~

by: (a) maintaining the minimum DNBR in the core 21.30 during normal opera-tion and during short term transients, (b) maintaining the peak linear power density 518.4 kW/f t during normal operation, and (c) maintaining the peak power density 5 20.4 kW/f t during short term transients. In addition, the above criteria must be met in order to meet the assu=ptions used for the loss-of-coolant accidents.

The power imbalance envelope defined in Figures 3.2-1 and 3.2-2 and the insertion limit curves, Figures 3.1-2 and 3.1-3 are based on LOCA analyses which have defined the maximum linear heat rate such that the maximum clad temperature will not exceed the Final Acceptance Criteria of 2200'F following a LOCA. Operation outside of the power imbalance envelope alone does not con-stitute a situation that would cause the Final Acceptance Criteria to be ex-ceeded should a LOCA occur. The power imbalance envelope represents the bound-ary of operation limited by the Final Acceptance Criteria only if the control rods are at the insertion limits, as defined by Figures 3.1-2 and 3.1-3 and if l the steady-state limit quadrant power tilt exists. Additional conservatism is introduced by application of:

a. Nuclear uncertainty factors.

b. Thereal calibration uncertainty,

c. Fuel densification effects.

d. Hot rod manufacturing tolerance factors.

e. Fotential fuel rod bow effects.

The action statements which permit limited variations from the basic require-ments are acce=panied by additional restrictions which ensures that the orig-inal criteria are met.

The definitions of the design limit nuclear power peaking factors as used in these specifications are as follows:

F Nuclear heat flux hot channel factor, is defined as the maximum local fuel 9 rod linear power density divided by the average fuel rod linear power den-sity, assuming nominal fuel pellet and rod dimensions.

DAVIS-BESSE, UNIT 1 B 3/4 2-1 8-16

POWER DISTRIBUTION LIMITS BASES

b. N The measurement of enthalpy rise hot channal factor, FSH, shall be in-creased by 5 percent to account for measurement error.

, For Condition II events, the core is protected from exceeding 20.4 kW/ft local-ly, and from going below a mini =um DNBR of 1.30, by automatic protection on power, axial power imbalance, pressure and temperature. Only conditions 1 l through 3, above, are mandatory since the axial power i= balance is an explicit input to the reactor protection system.

The quadrant power tilt limit assures that the radial power distribution sat-isfies the design values used in the power capability analysis. Radial power distribution measureuents are made during startup testing and periodically dur-r ing power operation.

The quadrant power tilt limit at which corrective action is required provides DNB and linear heat generation rate protection with x-y plane power tilts. In the event the tilt is not corrected, the margi.. for uncertainty on Fq is rein-stated by reducing the power by 2 percent for each percent of titt in excess of the limit.

3/4.2.5. DNB PARAMETERS The limits on the DNB related parameters assure that each of the parareters are =aintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the FSAR initial assumptions and have been analytically demonstrated adequate to main-tain a mini =um DNBR of 1.30 throughout each analyzed transient.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instru=ent read-out is sufficient to ensure that the parameters are restored within their lim-

_ its following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate using delta P instrumen-tation is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis, i~

L 7_

l DAVIS-BESSE, UNIT 1 B 3/4 2-3 8-17

3/4.4. REACTOR COOLANT SYSTDI BASES _ _ _

_5--

3/4.4.1. REACTOR COOLANT LOOPS The plant is designed to operate with both reactor coolant loops in operation, and maintain DNBR above 1.30 during all normal operations and anticipated transients. With one reactor coolant pump not in operation in one loop, thermal power is restricted by the nuclear overpower based on RCS flow and axial power imbalance and the nuclear overpower based on pump monitors trip, ensuring that the DNBR will be maintained above 1.30 at the maximum possible l thermal power for the number of reactor coolant pumps in operation or the lo-cal quality at the point of mir imum DNBR equal to 22%, whichever is more re-strictive.

A single reactor coolant loop provides sufficient heat removal capability for removing core decay heat while in hot standby; however, single f ailure con-siderations require placing a DHR loop into operation in the shutdown cooling mode if component repairs and/or corrective actions cannot be made within the allowable out-of-service ti=e.

3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pres-surized above its safety limit of 2750 psig. Each safety valve is designed to relieve 336,000 lbs per hour of saturated steam at the valve's setpoint.

The relief capacity of a single safety valve is adequate to relieve any over-pressure condition which could occur during shutdown. In the event that no safety valves are operable, an operating DER loop, connected to the RCS, pro-vides overpressure relief capability and will prevent RCS overpressurization.

During operation, all pressurizer code safety valves cust be operable to pre-vent the RCS from being pressurized above its safety limit of 2750 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from any transient.

Demonstration of the safety valves' lif t settings will occur only during shut-down and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

DAVIS-BESSE, UNIT 1 B 3/4 4-1 8-18

Figure B-1. Reactor Core Safety Limit 2tX)0 -

RC High Pressure Trip 619.0,2300 9

RC High 2200 _ Temperature Trip 619.0,21140 ACCEPTABLE e OPERATION Safety Limit 3

c.

RC Pressure E Temperature Trip E 2000 -

606.7,1985 RC tow Pressure Trip r

i

,_ 1800 __

i t i i I 590 600 610 620 630 Reactor Outlet Temperature, F TECHNICAL SPEClFICATION FIGURE 2. I-l Reactor Core Safety Limit I

L>

I L.

DAVIS-BESSE, UNIT 1 2-2 8-19 Babcock & Wilcox

Figure S-2. Reactor Core Safety Limit

% Rated Thermal Power

__ l 20

-35.8,112.0 4 PUMP LIMIT 22.4,112.0

-r 100

-35.8,86.7 3 PUNP LlHIT +22' 4

-60.0,80.0 48.0.80.0

-60.0,60.0 ( -- 60 48.0,60.0 ACCEPTABLE OPERATION

~~

FOR SPECIFIED RC PUMP COMBINATION

__ 20 UNACCEPTABLE UNACCEPTABLE OP ERATION OPERATION t i t , t ,

f

-60 -40 -20 0 20 40 60 Axial Power imbalance, %

TECHNICAL SPECIFI CATION FIGURE 2.1-2 Reactor Core Safety Limit PUMPS OPERATlHG REACTORCOOLANTFLOW(GPM) 4 337,200 3 290,l00 DAVIS-BESSE, UNIT 1 2-3 8-20 Babcock & \Vilcox

Figure 8-3. Trip Setpoint for Flux-4 Flux / Flow

". Rated Thermal Power n

120

-25.0,107.0 13.0,107.0 FOUR PUMP LIMIT LINES ..100 t

-40.0,84.0 -25.0,80.2 -

\

80.I -

curve shows t-ip 35.0,76.0 se tpoin t for a

'~ 25% flow reduction THREE PUNP for three pump LlHIT LINES operation. The

~60 actual trip setpc l - 40. 0, 57.2( will be calculate

'- by the RPS and wi

)35.0,49.2 be directly propo I~ ' tional to the gg actual flow with three pumps.

I

__20 ACCEPTABLE OPERATION FOR UNACCEPTABLE UNACCEPTABLE SPECIFIED RC PUMP COMBINATION OPERATION r -- OPERATION t ' '

~

7 ' '

f

-60 -40 -20 0 20 40 60 Axial Power imbalance, 7 TECHNICAL SPECI FICATION FIGURE 2. 2-1 Trip Setpoint for Fluxt Flux /F'ow DAVIS-BESSE, UNIT 1 2-7 Babcock & Wilcox 8-21

Figure 8-4. Allowable Value for Flux-a Flux / Flow

7. Rated Thermal Power

__ l20

-25.3,107.I25 13.3,(07.I25 FOUR PUMP LIMIT LINES-- 100 13.3,

-40.5,84.0 -25.3,80.325 0.325 80 THREE PUMP O 35.5,76.1 LIN8T LINES

~~

-40.5,57.2 <

) 35.5,49.3

.. 40

__ 20 ,

UNACCEPTABLE ACCEPTABLE UPERATION FOR UNACCEPTABLE OPERATION SPECIFIED RC PUMP COMBINATION OPERATION I f f f f f '

-60 -40 -20 0 20 40 60 Axial Power imbalance, f.

TECHNICAL SPECIFICATION FIGURE 2.2-2 Allowable value for Flux-A Flux / Flow DAVIS-BESSE, UNIT 1 2-8 Babcock & Wilcox 8-22 ,

1 Figure 8-5. Pressure / Temperature Limits d

2400 -

1 2300 -

^

oo O

{ 2200 -

0 S

o 5 2100 -

E o

E-( 2000 -

1900 -

I 1800 -

1700 ' ' ' ' '  ;

580 590 600 610 620 630 Reactor Outlet Temperature-( F)

RC FLOW POWER 387,200 GPM 112%

Pressure /Tc=perature Limits at Maximum Allowable Power for Minimum DNBR TECHNICAL SPECIFICATION BASES . FIGURE 2.1 DAVIS-BESSE, UNIT 1 B 2-8 8-23 Babcock & Wilcox

Figure 8-6. Regulating Group Position Limits, O to 150 10 EFPD, Four RCPs - Davis Besse 1, Cycle 2 UNACCEPTABLE OPERAll0N (281.102) (300.102) 100 - '

(235.100)

(274.92)

O E " ~

SHUTOCIN LIMli j (for 1% ak/k g shutdown margin) OPERAll0N

$ RESTRICIED

= 60 -

E

% (122.50) (225.50)

Y 3 ACCEPTABLE 2

- OPERAil0N t

5 20 - (26.15) 0 i '

O 100 200 300 GR 5 , , Roa Inaen (eitnarasn) g 0 75 100

$ GR 6 i ' ' '

8 0 25 75 100 x

g. '

GR 7 ' '

q 0 25 100

=:

o x

TECHNICAL SPECIFICATION FICURE 3.I-2a Regulating Group Position Limits, O to 150 _+ 10 EFPD, Four RCPs - Davis-Besse 1, Cycle 2 DAVIS-BESSE, UNIT 1 3/4 1-28

Figure 8-7. Regulating Group Position Limits, After 150 il9 EFPD,

• Four RCPs - Davis Bec- '., Cycle 2 (280.5.102) 100 -

(260 ICO) c 3(300.102)

(274,92' UNACCEPTABLE OPERATION O 80 -

(252.80) 2 SHUT 00XN LIMIT

' (for 1% Ak/k PERAil0N E shutdown margin) RESTRICTED

[ 60 -

E j (150.50) (225,50)

Y E j  % 40 -

ACCEPTABLE OPERATION b

E 20 -

( 'I I (90,15) 0 I '

0 100 200 300

$ CR 5 ' .

Rod index (witnarawn) o 0 75 100 Q- GR 6' ' ' '

e 0 25 75 100

$_ GR 7 i '

8 0 25 100 x

TECHNICAL SPECIFICATION ilGURE 3.1-2 b Regulating Group Position Limits, Af ter 150 110 EFPD, Four RCPs - Davis-Besse 1, Cycle 2 DAVIS-BESSE, UNIT 1 3/4 1-28a

Figure 8-8. Regulating Group Position Limits, O to 150 10 EFPD, Three RCPs - Davis Besse 1, Cycle 2 100 -

UNACCEPTABLE OPERATION U 80 -

j (184,77) (281,77) ,(300,77) 5 SHUT 00NN LIKIT (274,69.5) 3 (for 1% Ak/k shutclow margin (252.60.5)

[ 60 -

3

", (122,50) OPERAil0N RESTRICTED C

y { 40 -

(225,38)

$ cI" ACCEPTABLE OPERAT10N f

a.

20 -

(26,15)

(90,11.75) 0 t '

O 100 200 300 GR 5 ' Roa inden (witnarawn) g 0 75 100 8' GR 6 ' ' ' '

8 0 25 75 100 x-P CR 7 e i i h 0 15 100 8 TECHNICAL SPECIFICATION FIGURE 3.1-3a Regulating ';roup Position Limits, O to 150 *10 EFPD, x

Three RCPs - Davis-Besse 1, Cycle 2 DAVIS-BECSE, UNIT 1 3/4 1-29

1 , . I J Figure 8-9. Regulating Group Position Limits, After 150 !10 LFPD, Three RCPs - Davis Besse 1, Cycle 2 100 -

S. 80' -

UNACCEPTABLE OPERATION (207.77) (283.5,77)  ? (300,77) n.

SHUTOORN LinflT l (274,69.5) 5 (for 17.ak/k

$ 60 -

shutdown margin) (252.60.5)

B >

E (150.5P)

/ OPERATION E 40 RESTRIOTE0 y y (225.38) 5 E 20 -

(35,15) ACCEPTABLE OPERAll0N (90.11.75) 0 ' '

0 100 200 300 GR Si i bod Index (withdrann) m 0 75 100 n>

g CR E i i '

o 0 25 15 100 x-e GR Ti ' '

G 0 25 100 W

R TECHNICAL SPECIFICAil0N FIGURE 3.1-3b Regulating Group Position Limits, After 150 110 EfPD, Three RCPs - Davis-Besse 1, Cycle 2 DAVIS-BESSE, UNIT 1 3/4 1-29a

Figure 8-10. APSI. Position Limits, O to 150 :10 EFPD, Four RCPs - Davis-Besse 1, Cycle 2 (12,102) (30,102) 100 -

,(8,92) (35,$2)

C

$ 80 g(8,80) (40,80) a.

j RESTRICTED REGION 5

g 60 -

(64,50) d((5,50)

(100,45) o E 40 -

5

$ PERMIS318LE 20 -

REGION

(0,0) 0 20 40 60 80 100 APSR Position (% witndrawn)

TECHNICAL SPECIFICATION FIGURE 3.1-5a APSR Position Limits, O to 150 i10 EFPD, Four RCPs - Davis-Besse 1, Cycle 2 DAVIS-BESSE, UNIT 1 3/4 1-35 8-28 Babcock & Wilcox

Figure 8-11. APSR Position Limits, After 150 :10 EFFD, Four RCPs - Davis Besse 1, Cycle 2 (12.102) (30,102) 100 i

7 (8,92) (35,92)

E

80 (5,80) (40,80)

RESTRICTED

~

5 REGION e

B E 60 O

(64,50) 5 (g(5,50)

S ' N (100,45) 40 -

t g PERMISSIBLE OPERATING REGION i

20 -

t t

(0,0) , ,

0 20 40 60 80 100 APSR Position (", Withdrawn)

, TECHNICAL SPECIFICATION FIGURE 3.1-5 b APSR Position Limits, Af ter 150 *10 EFPD, Four RCPs - Davis-Besse 1, Cycle 2 DAVIS-BESSE, UNIT 1 3/4 1-36 8-29 Babcock t. Wilcox

Figure 8-12. APSR Position Limits, 0 to 150 10 EFPD, Three RCPs - Davis Besse 1, Cycle 2 100 -

j 80 - (12,7 7) (30,77) f 5 (8,69.5) (35,69.5) j RESTRICTE0 (40,60.5)

[ 60 -

9(8,60.5) REGION 3 J E f E

e 40 (64,38) y (5,38) (100,34.25) b a -

E PERMISSIBLE 20 -

OPERATING REGION

( 0,0 )

0 ' ' ' I O 20 40 60 80 100 APSR Positian (% Withdrawn)

TECHNICAL SPECIFICATION FIGURE 3.1-5 c APSR Position Limits, O to 150 *10 EFPD, Three RCPs - Davis-Besse 1, Cycle 2 DAVIS-BESSE, UNIT 1 3/4 1-37 8-30 Babcock & Wilcox

Figure 8-13. APSR Position Li=its, After 150 :10 EFFD, Three RCPs - Davis Besse 1, Cycle 2 100 -

m 3 80 - (12,77) (30,77) d'

~

U (8.69.5) (35,69.5) a (5,60.5) (40,60.5)

.- 60 - RESTRICTED m

i g REGION m

I E E 40 - (64,38)

S (100,34.25)

(f (5,38) -O E'

v U

J' PERMISSIBLE 20 -

OPERATING REGION (0,0) 0 ' ' ' 1 0 20 40 60 80 100 APSR Position (% Witndrawn) l TECHNICAL SPECIFICATION FIGURE 3.1 -5 d APSR Position Limits, Af ter 150 10 EFPD, i Three RCPs - Davis-Besse 1, Cycle 2 L

DAVIS-BESSE, UNIT 1 3/4 1-38 8-31 Babcock & Wilcox

Figure 8-14. Axial Power Imbalance Limits, O to 150 210 EFPD, Four RCPs - Davis Besse 1, Cycle 2

(-14.5.102) ,,

(+10,102 )

( - 18,92 ) (+12,92) n

( -2 5. 5,8 0) 80 -- 0 (+20,80)

E o.

E 60 -- 5, 3m

(-30,50) [C (+25,50) 40 -- g PERMISSIBLE $

OPERATING REGION

[

f

s 1 1

-30 -20 -10 0 +10 +20 +30 Axial Power imoalance (5)

TECHNICAL SPECIFICATION FIGURE 3.2 la Axial Power Imbalance Limits, O to 150 10 EFPD, Four RCPs - Davis-Besse 1, Cycle 2 DAVIS-BESSE, UNIT 1 3/4 2-2 8-32 Babcock & Wilcox

'~ . .

Figure 8-15. Axial Power Imbalance Limits, After 150 !10 EFPD, Four RCPs - Davis Besse 1, Cycle 2

(-18.2.102) o (+15,102 )

(-20,92) (+15,92) RESTRICTED REGION

(-27,80) (+20,80) 80 --3 E

m E

60--

_ g

(-35,50) = (+25,50)

E ee 40--5 S

E O

PERMISSIBLE 20--

l OPERATING REGION i , i i .

-30 -20 -10 0 +10 +20 +30 Axial Power ImDalance (%)

TECHNICAL SPECIFICATION FIGURE 3.2-lb Axial Power Imbalance Limits, Af ter 150 10 EFPD, Four RCPs - Davis-Besse 1, Cycle 2

(

DAVIS-BESSE, UNIT 1 3/4 2-2a 8-33 Babcock & Wilcox

Figure 8-16. Axial Power Imbalance Limits, O to 150 110 EFPD, Three RCPs -- Davis Besse 1, Cycle 2 100 --

80 - -

RESTRICTED

( -10. 9,7 7 ) (+7.5,77)

REGION

( - 12, 69. 5 )

l[ (+9.69.5) 5 o.

( - 19.1. 60. 5 ) 60 --  ; +15,60.5)

E E

E

( -22. 5,38) 40 -- *'

.- (+18.75,38)

C Z

8 a

o_

PERMISSIBLE 20 -- }{m OPERATING E o.

REGION i i t i i i

-30 -20 -10 0 +10 +20 +30 Axial Power Ir.calance (%)

TECHNICAL SPECIFICATION FIGURE 3.2-2a Axial Power Imbalance Limits, O to 150 Il0 EFFD, Three RCPs - Davis-Besse 1, Cycle 2 DAVIS-BESSE, UNIT 1 3/4 2-3 8-34 Babcock & Wilcox

Figure 8-17. Axial Power Imbalance Limits, After 150 10 EFPD, Three RCPs - Davis Besse 1, Cycle 2

7. of Rated Thermal Power

. 100 RESTRICTED REGION

(-13.7,77) -_ 80 (11.3,77)

F

(- 15,6 9. 5 ) (l1.3,69.5)

(-20.3,60.5) y -- 60 (l5,60.5) i

(-26.3,38) (18.75,38)

_ t;o PERHISSIBLE OPERATlHG f REGION .

! -- 20 t

I e t I t I

-30 -20 -10 0 10 20 30 Axial Power imbalance, 7 TECHNI CAL SPECI FI CATION FIGURE 3.2-2b Axial Power Imbalance Limits, After 150 10 EFPD,

'Ihree RCPs - Davis-Besse 1, Cycle 2 I

m DAVIS-BESSE, UNIT 1 3/4 2-3a

~

8-35 Babcock & Wilcox

Figure 8-18. ., Control Rod Core Locati'ons and Group Assign::lents -

Davis-Besse 1, Cycle 2 I

I A

B 4 1 7 4 C

1 6 6 1 D 7 8 2 8 7 E

1 5 5 1

, F 4 8 7 5 7 8 4 C 6 3 3 6 V- 7 2 g 5 3 5 2 7 -Y K 6 3 3 6

1. 4 8 7 5 7 8 4 .

.f M 1 5 5  !

1 N 7 8 2 8 7 0 1 6 6 1 P 4 7 4 P.

I z

1 2 3 4 5 6 7 8 9 10 11 '12 13 14 15 No. of

_ Croup rods Purpose, 1 8 Safety 2 4 Safety 3 5 Safety 4 8 Safety

'5 8 Control 6 8 Control 7 12 control 8 ,_8 APSRs Total f 61 Technical Specification Figure 3.1-4. Control Rod Core Locations and Grotp Assignmen N Davis-Besse 1, Cycle 2 DAVIS-BESSE, UNIT 1 3/4 1-31

9. STARTUP PROGRAM - PHYSICS TESTING The planned startup test program associated with core performance is outlined below. These tests verify that core perfor=ance is within the assumptions of

~

the safety analysis and provide confirmation for centinued safe operation of the unit.

9.1. Precritical Tests 9.1.1. Control Rod Trip Test Precritical control rod drop times are recorded for all control rods at hot full-flow conditions before zero power physics testing begins. Acceptable criteria state that the rod drop time from fully withdrawn to 75% inserted shall be less than 1.66 seconds at the conditions above, It should be noted that safety analysis calculations are based on a rod drop f

time of 1.40 seconds from fully withdrawn to two-thirds inserted. Since the most accurate position indication is obtained from the zone reference switch at the 75% inserted position, this position is used instead of the two-thirds inserted position for data gathering. The acceptance criterion of 1.40 seconds corrected to a 75% inserted position (by rod insertion versus time correlation) is 1.66 seconds.

9.1.2. Reactor Coolant Flow Reactor coolant (RC) flow with four reactor coolant pumps (RCPs) running will be measured at hot zero power (HZP) steady-state conditions. Acceptance cri-teria require that the measured flow be within allowable limits.

9.1.3. RC Flow Coastdown

. The coastdown of RC flow from the tripping of the highest flow RCP from four RCPs running vill be measured at HZP conditions. The coastdown of RC flow versus time will then be compared to the required RC flow versus time to de-termine if acceptance is met.

9-1 Babcock & VVilcox

9.2. Zero Power Physics Tests 9.2.1. Critical Baron Concentration Criticality is obtained by deboration at a constant dilution rate. Once crit-icality is achieved, equilibrium boron is obtained and the critical boron con-centration determined. The critical baron concentration is calculated by cor-recting for any rod withdrawal required in achieving equilibrium baron. The acceptance criterion placed on critical boron concentration is that the actual boron concentration must be within 100 ppm boron of the predicted value.

9.2.2. Temperature Reactivity Coefficient The isothernal temperature coefficient is measured at approximately the all-rods-cut configuration and at the HZP rod insertion limit. The average cool-ant temperature is varied by first decreasing then increasing the temperature by 5*F. During the change in temperature, reactivity feedback is compensated by discrete change in rod motion; the change in reactivity is then calculated by the su=mation of reactivity (obtained from reactivity calculation on a strip chart recorder) associated with the temperature change. Acceptance criteria state that the measured value shall not differ from the predicted value by more than 0.4 x 10-4 (ak/k)/*F (predicted value obtained from Physics Test Manual curves).

The moderator coefficient of reactivity is calculated in conjunction with the temperature coefficient measurement. After the temperature coefficient has been measured, a predicted value of fuel Doppler coef ficient of reactivity is added to obtain moderator coefficient. This value must not be in excess of the acceptance criteria limit of +0.9 x 10-4 (ak/k)/*F.

9.2.3. Control Rod Group Reactivity Worth Control bank group reactivity worths (groups 5, 6, and 7) are measured at HZP conditions using the boron / rod swap method. The boron / rod swap method con-sists of establishing a deboration rate in the RC system and compensating for the reactivity changes of this deboration by inserting control rod groups 7, 6, and 5 incremental steps. The reactivity changes that occur during these measurements are calculated based on reactimeter data, and differential rod ,

worths are obtained from the measured reactivity worth versus the change in rod group position. The differential rod worths of each of the controlling 9-2 Babcock & \Vilcox

groups are then summed to obtain integral rod group worths. The acceptance criteria for the control bank group worths are as follows:

1. Individual bank 5, 6, 7 vorth:

predicted value - measured value x 100 s 15.

measured value

- 2. Sum of groups 5, 6, and 7:

predicted value - measured value measured value 100 $ 10.

9.2.4. Ejected Control Rod Reactivity Worth r- After CRA groups 7, 6, and 5 have been positioned near the minimum rod inser-tion limit, the ejected rod is borated to 100% withdrawn and the worth obtained

, by adding the incremental changes in reactivity by boration.

Af ter the ejected rod has been borated to 100% withdrawn and equilibrium boron established, tha ejected rod is then swapped in versus the controlling rod group and the worth determined by the change in the previously calibrated con-trolling rod group position. The boron and rod swap values are averaged and error-adjusted to determine ejected rod worth. Acceptance criteria for the ejected rod worth test are as follovs:

1. predicted value - measured value x 100 s 20.

measured value r

2. Measured value (error adjusted) s 1.0% Ak/k.

The predicted ejected rod worth is given in the Physics Test Manual.

i 9.3. Power Escalation Tests

( 9.3.1. Core Power Distribution Verification at 40,t75, and o 100% FP With Nominal Control Rod Position Core power distribution tests are performed at N40, N75, and N100 full power (FP). The test at 40% FP is essentially a check on pcwer distribution in the core to identify any abnormalities before escalating to the 75% FP plateau.

Rod index is established at a noninal FP rod configuration at which the core power distribution was calculated. APSR position is established to provide a core power imbalance corresponding to the imbalance at which the core power distribution calculations were performed.

9-3 Babcock & Wilcox

The following acceptance criteria are placed on the 40% FP test:

1. The worst-case maximum linear heat rate must be less than the LOCA limit.

2. The minimum DNBR must be greater than 1.30.

3. The value obtained from the extrapolation of the minimum DNBR to the next power plateau overpower trip setpoint must be greater than 1.30 or the ex-trapolated value of imbalance must fall outside the RPS power / imbalance /

flow trip envelope.

4. The value obtained from the extrapolation of the worst-case maximum linear heat rate to the next power plateau overpower trip setpoinc must be less than the fuel melt limit or the extrapolated value of imbalance must fall outside the RPS power / imbalance / flow trip envelope.

5. The quadrant power tilt shall not exceed the limits specified in the Tech-nical Specifications.

6. The highest measured and predicted radial peaks shall be within the fol-lowiug limits:

predicted value - measured value '

measured value 100 5 8.

7. The highest measured and predicted total peaks shall be within the follow-ing limits:

predicted value - measured valu 100 < 12.

measured value Items 1, 2, 5, 6, and 7 above are established to verify core nuclear and ther-mal calculational models, thereby verifying the acceptability of data from these models for input to safety evaluations.

Items 3 and 4 establish the criteria whereby escalation to the next power pla-teau may be accomplished without exceeding the safety limits specified by the safety analysis with regard to DNBR and linear heat rate.

The power distribution tests performed at 75 and 100% FP are identical to the 40% FP test excent that core equilibrium xenon is established prior to the 75 and 100% FP tests. Accordingly, the 75 and 100% FP measured peak acceptance criteria are as follows:

9-4 Babcock & kVilcox

~

1. The highest measured and predicted radial peaks shall be within the follow-ing limits:

predicted value - measured value measured value x 100 s 5.

2. The highest measured and predicted total peaks shall be within the follow-ing limits:

predicted value - measured value x 100 $ 7.5.

measured value

-- 9.3.2. Incore Vs Excore Detector Imbalance Correlation Verification at N40% FP Imbalances are set up in the core by control rod positioning. Imbalances are read simultaneov.ly on the incore detectors and excore power range detectors for various icbalances. The excore versu's incore detector' offset slopes must be at least 1.15. If the excore versus incore detector offset slope crite-rion is not =et, gain amplifiers on the excore detector signal processing equipment are adjusted to provt'e the required gain.

9.3.3. Temperature Reactivity Coefficient at s100% FP The average RC temperature is decreased and then increased by about 5'F at con-stant reactor power. The reactivity associated with'each temperature change is obtained from the change in the controlling rod group position. Controlling rod group worth is measured by the fast insert / withdraw cethod. The tempera-ture reactivity coefficient is calculated from the measured changes in reac-tivity and temperature.

Acceptance criteria state that the moderator temperature coefficient shall be negative.

9.3.4. Power Doppler Reactivity Coefficient at 4100% FP Reactor power is decreased and then increased by abo'ut 5% FP. The reactivity change is obtained from the change in controlling rod group position. Control

- rod group worth is measured using the fast insert / withdraw method. Reactivity corrections are made for changes in xenon and RC temperature that occur during the measurement. The power Doppler reactivity coefficient is calculated from tpe measured reactivity change, adjusted as stated above, and the measured power change.

9-5 Babcock & \Vilcox

The predicted value of the power Doppler reactivity coefficient is given in the Physics Test Manual. Acceptance criteria state that the =casured value shall be more negative than -0.55 x 10-4 (ak/i)/% FP.

9.4. Procedure for Use When Acceptance Criteria Are Not Met' _ . _ _ _. _

If acceptance criteria for any test are not met, an evaluation is performed with participation by Babcock & Wilcox technical personnel as required. Further specific actions depend on evaluation results. These actions can include repeating the tests with more detailed attention to test prerequisites, added tests to search for anomalies, or design personnel performing detailed analyses of potential safety problems because of parameter deviation. Power is not escalated until evaluation shows that plant safety will not be compromised by such escalation.

. ~

9-6 Babcock & \Vilcox

REFERENCES i

1 BPRA Retainer Design Report, BAW-1496, Babcock & Wilcox, Lynchburg, Virginia, May 1978.

2 Davis-Besse Unit 1, Final Safety Analysis Report, Docket No. 50-346.

3 Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse, BAW-10084, Rev 1, Babcock & Wilcox, Lynchburg, Virginia, Novem-I' ber 1976.

4 C. D. Morgan and H. S. Kao, TAFY -- Fuel Pin Temperature and Gas Pressure I Analysis, BAW-10044, Babcock & Wilcox, Lynchburg, Virginia, May 1972.

5 H. H. Hassan, W. A. Wittkopf, and W. H. Mullan, Babcock & Wilcox Version i' of PDQ07 - User's Manual, BAW-10ll7A, Babcock & Wilcox, Lynchburg, t

Virginia, January 1977.

6 Attachment 1 to Application to Amend Operating License for Removal of Burnable Poison Rod and Orifice Rod Assemblies, BAW-1489, Rev 1, Babcock i & Wilcox, Lynchburg, Virginia, May 1978.

7 Davis-Besse Unit 1 Fuel Densification Report, BAW-1401, Babcock & Wilcox, Lynchburg, Virginia, April 1975.

8 L. S. Rubenstern (USNRC) to J. H. Taylor (B&W), Letter, " Evaluation of I

Interim Procedure for Calculating DNBR Reductions Due to Rod Bow,"

t October 18, 1979.

l 9 W. L. Bloomfield, et al. , ECCS Evaluation of B&W's 177-FA Raised-Loop NSS, BAW-10105, Rev 1, Babcock & Wilcox, Lynchburg, Virginia, July 1975, 10 S. A. Varga (USNRC) to J. H. Taylor (B&W), Letter, " Update of BAW-10055,

' Fuel Densification Report,'" December 5, 1977.

A ~. Babcock & Wilcox

,[' .

() () Attachment 1 to BAW-1598, January,1980, File: 0017;0028 Revised by TEco, 2/4/80 r TCLECO Docket No. 50-346 b U[.333C3 License No. NPF-3 LOWELt E. ROE Serial No. 52 7 V.u Pres. cent Facasetnes Cowoomens July 13, 1979 mas ase sa a Director of Nuclear Reactor Regulation Attention: Mr. Robert N. Reid, Chief Operating Reactors Branch No. 4 Division of Operating Reactors United States Nuclear Regulatory Commission Washington, .D. C. 20555 .

Dear Mr. Reid:

Under separate cover,' we are transmitting ihree (3) original and forty (40) conformed copies of an application for Amendment to Facility Operating License No. NPF-3 for the Davis-Besse Nuclear Fower Station Unit No. 1.

This application requests that the Davis-Besse Nuclear Power Station Unit 1 Technical Specification, Appendix A,be revised to reflect the changes attached.

The proposed changes include 1) addition of requirements on the interim antici-patery reactor trip system in sections 3.3.2.3, 4.3.2.3 and Table 3. 3-15 and Tatie 4.3-15; 2) addition of require =ents for auxiliary feedwater flow indica-tians in section 4.7.1.2; 3) change to ceactor coolant high pressure reactor trip setpoint.in Table 2.2-1; 4) addition of a reactor coolant pressure temperature curve, Figure' 3.4-5; 5) addition of require =ents in sections 3.4.3 and 4.4.3 for the electromatic relief valve; 6) change to technical specificatien basis ,

2.2.1; and 7) change to SFAS actuation mode and actions of Tables 3.3-3 and 4.3-2.

Thir, amendment request involves several changes of Class III type. It is there-fore dater =ined to be a Class IV a~endment. Enclosed is $12,300.00 as required by 10CFR170.

The seven attach =ents identify each proposed change, its safety evaluation and schedule require'd to implement the change after NRC approval. Ite=s 1-5 above fulfill the seven day require =ents of your letter of July 6, 1979.

. Yours - ry truly,

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THE TCLEGO ECISCN COMPANY ED!SCN PLAZA 300 MAC!SCN AVENUE TOLEDO. CHIO 43S52

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APPLICATION FOR AMENDMENT TO FACILITY OPERATING LICENSE NO. NPF-3 FOR DAVIS-BESSE NUCLEAR POWER STATION UNIT NO. I s

Enclosed are forty-three (43) copies of the requested changes to the Davis-Besse Nuclear Power Station Unit No. 1 Technical Specifications, Appendix A to Facility Operating License No. NPF-3, together with the Safety Evaluation for the requested change. The proposed changes include

1) addition of requirements on the interim anticipatory reactor trip system in sections 3. 3. 2. 3, 4. 3. 2. 3 and Table 3.3-15 and Table 4.3-15; 2) addition of requirements for auxiliary feedwater flow indications in section 4.7.1.2;

3) change to reactor coolan_t high pressure reactor trip setpoint in Table 2.2-1;

4) addition of a reactor coolant pressure temperature curve, Figure 3.4-5;

5) addition of requirements in sections 3.4.3 and 4.4.3 for the electromatic relief valve; 6) change to technical specification basis 2.2.1; and 7) change to SFAS actuation mode and actions of Tables 3.3-3 and 4.3-2.

By s/ Lowell E. Roe Vice President, Facilities Development

. s Sworn to and subscribed before me this . thirteenth day of .Iuly,1979.

u. /s O . o Y 7 Notary Public LINDA L COSTELt?

Notary Pubhc - State of Ohio My Commission Expires Feb. 9,1982

'jocket No. 50-346 ., .-

License No. NPF-3 Serial No. 527 July 13,1979 Page One of Two III Changes to Davis-Besse Nuclear Power Station, Unit 1 Technical Specifica-tions. Appendix A, Table 2.2-1 concerning the reactor coolant high pressure reactor protection system trip setpoint.

A. Time Required to Implement This change can be effective upon NRC issuance B. Reason for Change (Facility Change Request 79-170)

The reactor coolant high pressure reactor protection system trip setpoint was reduced in order to decrease the number of transients that could open the pressurizer pilot operated relief valve. This change reduces the upper bound of the current technical specification to be consistent with the recently revised setpoint.

C. Safety Evaluation ,

This enanse provides for lowering the Reactor Coolant System (RCS) high pressure trip setpoint from 2351.4 psig to 2296.4 psig in the Reactor Protecticn System (RPS).

From the calculations performed by B&W it is evident that if the RPS high pressure trip setpoint is changed from 2351.4 psig to 2296.4 psig the peak pressurizer pressure will remain below 2345 psig during loss of feedwater transients. The severity of these transients is reduced by implementing the change proposed by this FCR as described below.

In the case of loss of main feedwater the Steam and Feedwater Rupture Control Syste= (SFRCS) isolates both steam generators (SG) on the feed-water and steam side, and auxiliary feelwater is initiated to the SGs.

Because of the insufficient heat tran3fer to the seccndary side of the SGs caused by loss of feedwater, the reactor is tripped on high RCS pressure.

It is estimated that with the present RPS high pressure trip setpoint the reactor will trip aporoximately 20 seconds af ter a loss of feedwater from 100% power as compared to 13 seconds with the lower RPS high pressure trip setpoint Therefore, with the lower setpoint there is less energy (corresponding; to the seven second time difference) contained in the RCS.

On the s acondary side the steam pressure builds up because of main steam line isolation and the code safety valves lift (on the secondary side) to relieve the pressure. Since the energy ' content of the RCS is lower for the lower RPS trip setpoint, the secondary side code safety valves lif t for a shorter duration. Consequently, when the secondary system code safety valves reseat, the secondary side level in the steam generators is higher for the lower RPS trip setpoint as compared to that which would be obtained with the present (higher) RPS trip setpoint. Also, when -the level in the steam generator is brought to normal through auxiliary feedwater operatien for the case of the lower RPS trip setpoint, a smaller amount of cold auxi-liary feedwater will be added to the steam generators, thereby - t'ucing the net volu=etric contraction of the RCS. .

, s .,

Docket No. 50-346 License No. NPF-3 Serial No. 527 July 13, 1979 Page Two of Two C. Safety Evaluation (Continued)

This will result in a higher minimum pressurizer 1 "el during such a transient with the lower RPS trip setpoint.

It should be noted that the RCS temperature and pressure remain the same after the secondary system code safety valves have reseated regardless of the hip,her or lower RPS high pressure trip setpoint.

This assumes that the pressurizer electromatic relief valve does not actuar during the transient. With the increase in the relief setpoint of the electromatic relief valve to 2400 psig, actuation of the relief valve will be eliminated during a loss of feedwater transient. Also, with the increase in electromatir. relief valve actuation setpoint the amount of reactor coolant lost to .he pret e eizer quench tank on lif ting of the electromatic. relief valva will be eliminated, leading to higher reactor coolant inventory in thr: RCS,to maintain a higher pressurizer level.

In su= mary, the proposed reduction in the RPS high pressure trip setpoint does not degrade the safety of the plant and does not invalidate any of the safety analyses pcesented in the Davis-Besse Unit 1 FSAR or in the safety evaluation submitted to the NRC on December 22, 1978 (Serial Nc. 475). The posssibility of an accident or a malfunction of a different type than any evaluated in the FSAR is not created. Also, the margin of safety as defined in the bases for technical specification is not reduced.

Pursuant to the above, the proposed change does not involve an unreviewed safety question.

Because 2300 psig in Table 2.2-1 is less than existing trip setpoint in license of 2355 psig, the setpoint change can be made prior to receiving the license snend=ent.

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